1. Technical Field
This invention relates to a class of devices for creating a range of corrosive conditions simulating those found in crevices of metal surfaces contained within a corrosive medium. It has particular application to the conditions found in heat exchangers, especially at the surfaces of heat exchange tubes within the secondary side of pressurized water reactor nuclear steam generators.
This invention additionally relates to methods of monitoring and predicting tube crevice corrosion, also having applications to the operation of heat exchangers.
2. Description of the Prior Art
The thin-walled metal tubes that are generally used in heat exchangers often suffer highly stressed conditions during extended periods of use. Stresses caused by elevated temperatures and pressures in combination with a corrosive chemical environment can lead to failure of these metals. This process occurs frequently at the crevices formed at interfaces with supporting structures. Corrosive debris suspended in a heat exchanger fluid can accumulate in the crevices as a corrosive sludge. Where a heat exchanger contains a hazardous material, or where the heat exchanger is within a system that is difficult or costly to shut down for maintenance, it is especially important to be able to monitor corrosive degradation, or even better, to predict catastrophic failure before it happens. Environmental concerns and cost considerations are both important in pressurized water reactor (PWR) nuclear steam generators.
Because corrosion of the heat exchanger tubes of PWR steam generators present special concerns related to this invention, it is worthwhile to outline some of the details of these systems. Nuclear powered steam generators have three principle parts. A primary side contains radioactive hot water heated by the nuclear core. A secondary side contains non-radioactive water which, upon being converted into steam, powers the turbine generator. Heat is transferred from the primary side to the secondary side by a heat exchanger comprising a tube sheet in which the inlet and outlet ends of a plurality of U-shaped tubes are mounted. In PWR's, water from the primary side enters the U-shaped tubes through the inlets, flows through the U-shaped tubes and exits the outlets which are hydraulically isolated from the inlets by a divider sheet. A second hydraulic flowpath circulates non-radioactive water around the outside surfaces of the U-shaped tubes extending into the secondary side. Heat from the primary side is transferred across the metal boundaries of the U-shaped tubes to the secondary side.
The U-shaped heat exchange tubes of nuclear powered steam generators are subjected to conditions which can lead to corrosion and failure of the tubes. Corrosion in the crevice regions of the heat exchanger is especially troublesome. Crevices are formed in the annular space between the heat exchange tubes and the tube sheet and also in the annular clearance between the tubes and the support plates in the secondary side. The support plates are used to uniformly space and align these tubes which otherwise would be buffeted about by the strong hydraulic flow around them. A sludge formed of particulates in suspension in the secondary side tends to collect in these crevices. These sludges are comprised predominantly of iron oxides and copper. The normal hydraulic circulation of the water in the secondary side is not sufficient to flush out the sludge from these crevices. In fact, the poor hydraulic circulation in these regions exacerbates the situation. The collection of sludge in the crevices impedes heat transfer from these regions, creating localized areas of elevated temperature (or "hot spots") in the tubes adjacent the sludge. The elevated temperatures in the "hot spots" allow higher local concentrations of corrosive impurities in solution, accelerating the corrosion process. Nuclear power plant operators periodically attempt to sweep the sludge out of the generator vessel by hydraulic means, however this is not always effective. The U-shaped heat exchange tubes of nuclear steam generators are typically formed from corrosion resistant nickel alloys, such as Inconel.RTM. 600, but corrosion in the crevices due to the elevated heat conditions and the high pressures found within the U-shaped tubes may ultimately penetrate the tube wall, resulting in leakage of radioactive water from the primary side into the non-radioactive water in the secondary side of the steam generator. Remedial action taken during maintenance shutdowns of the reactor can prevent such leakage, however it would be very useful to be able to know when corrosion is occurring before there is significant propagation.
While early forms of steam generator corrosion (thinning and denting) could be readily related to a particular operating chemistry, more recent forms of corrosion (intergranular corrosion and cold leg thinning) usually cannot be related to a direct cause. As impurity levels in steam generators have been reduced, it has become progressively more difficult to correlate contaminant levels, as measured either in the steam generator blowdown during power operation or by monitoring the hideout return following a shutdown, with the severity of corrosion, as measured by the time period required to initiate corrosion, its subsequent propagation rate, and the number of affected tubes. Because of the inability of using the operating chemistry as a means of judging the possible future extent of corrosion, a more direct means of relating a particular operating chemistry to the future occurrence of corrosion is highly desirable. With such a capability, it would be possible to undertake a corrective action well in advance of the corrosion actually occurring.
For this reason, model steam generators were developed to monitor the corrosion occurring in the heat exchange tubes of particular steam generators so that corrective action could be taken before failure of the tube walls. Model steam generators are bulky complicated apparatuses that attempt to mimic the actual steam generators. They contain all the hydraulic elements of an actual steam generator including circulating primary water, circulating secondary water and heat exchange tubes. They operate by subjecting an array of sample heat exchange tubes to a set of heat, pressure and chemical conditions approximating that which surrounds the heat exchange tubes in such nuclear steam generators. Their accuracy depends upon the accuracy of the simulation of the conditions within the operating steam generator. One type of model steam generator is described in U.S. Pat. Nos. 4,628,870, 4,635,589 and 4,640,233, all which were assigned to the Westinghouse Electric Corporation. A more compact model steam generator was disclosed in U.S. Pat. No. 4,637,346, also assigned to the Westinghouse Electric Corporation.
Model steam generators, while being useful instruments for simulating conditions in an actual steam generator, do not subject the tubes in the model steam generators to exactly the same conditions as found in an operating generator. Although the same secondary side feedwater source as used in a PWR generator may be fed into the model generator's secondary side, this does not guarantee the same chemistry as that found within the power plant's secondary side. An additional defect they suffer is that it is difficult to change the primary tubes in the model steam generators, for example, to do inspection or destructive testing of the heat exchange tubes.